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MatMCNP v. 4.0

RESOURCE

Publicly Accessible Repository
https://github.com/sandialabs/MatMCNP
https://doi.org/10.11578/dc.20190124.1

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Abstract

https://www.osti.gov/doecMatMCNP is a code for generating material cards for the MCNP radiation transport code. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.
Developers:
Depriest, Kendall [1] ;  Saavedra, Kevin
  1. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Release Date:
2019-01-16
Project Type:
Open Source, Publicly Available Repository
Software Type:
Scientific
Programming Languages:
Fortran
Python
Version:
v. 4.0
Licenses:
BSD 3-clause "New" or "Revised" License
Sponsoring Org.:
Code ID:
22907
Site Accession Number:
SCR 1716.1
Research Org.:
Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
Country of Origin:
United States

RESOURCE

Publicly Accessible Repository
https://github.com/sandialabs/MatMCNP
https://doi.org/10.11578/dc.20190124.1

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Citation Formats

Depriest, Kendall, and Saavedra, Kevin. MatMCNP v. 4.0. Computer Software. https://github.com/sandialabs/MatMCNP. USDOE. 16 Jan. 2019. Web. doi:10.11578/dc.20190124.1.
Depriest, Kendall, & Saavedra, Kevin. (2019, January 16). MatMCNP v. 4.0. [Computer software]. https://github.com/sandialabs/MatMCNP. https://doi.org/10.11578/dc.20190124.1.
Depriest, Kendall, and Saavedra, Kevin. "MatMCNP v. 4.0." Computer software. January 16, 2019. https://github.com/sandialabs/MatMCNP. https://doi.org/10.11578/dc.20190124.1.
@misc{ doecode_22907,
title = {MatMCNP v. 4.0},
author = {Depriest, Kendall and Saavedra, Kevin},
abstractNote = {https://www.osti.gov/doecMatMCNP is a code for generating material cards for the MCNP radiation transport code. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.},
doi = {10.11578/dc.20190124.1},
url = {https://doi.org/10.11578/dc.20190124.1},
howpublished = {[Computer Software] \url{https://doi.org/10.11578/dc.20190124.1}},
year = {2019},
month = {jan}
}